ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 21

Document Title:

Energy Technology Engineering Center – 

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0038-4 

Revision: 00

Effective Date: 03/01/2006

Type of Document: TBD

Supersedes: None 

Subject Expert: Tracy A. Ikenberry

Document Owner

Approval: Signature on File Approval Date: 02/10/2006

Melton H. Chew, TBD Team Leader

Approval: Signature on File Approval Date: 02/28/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 03/01/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/01/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 2 of 21

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/01/2006 00 New technical basis document for the Energy Technology 

Engineering Center – Occupational Environmental Dose. First 

approved issue. Training required: As determined by the Task 

Manager. Initiated by Melton H. Chew.
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 3 of 21

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

4

4.1 Introduction ..............................................................................................................................5

4.2 Operations Overview................................................................................................................6

4.3 Inhalation of Onsite Airborne Radionuclides............................................................................. 7

4.3.1 Stack Effluent Concentrations, 1988 to 1999................................................................8

4.3.2 Stack Effluent Concentrations, 1971 to 1987................................................................9

4.3.3 Stack Effluent Concentrations, 1959 to 1970................................................................9

4.3.4 Stack Effluent Concentrations, Before 1959 ................................................................. 9

4.3.5 Estimates of Potential Inhalation Intake, 1952 to 1999 ............................................... 10

4.4 Ambient External Radiation.................................................................................................... 12

4.4.1 External Dose, 1996 to 1999 ......................................................................................13

4.4.2 External Dose, 1990 to 1995 ......................................................................................13

4.4.3 External Dose, 1975 to 1989 ......................................................................................13

4.4.4 External Dose, 1959 to 1974 ......................................................................................14

4.4.5 External Dose Before 1959.........................................................................................14

4.4.6 Estimates of External Dose, 1952 to 1999..................................................................14

4.5 Ingestion ................................................................................................................................16

4.6 Uncertainty.............................................................................................................................17

References ........................................................................................................................................18

Glossary ............................................................................................................................................21

LIST OF TABLES

TABLE TITLE

4-1 Identified radionuclides in facility air effluents, 1988 to 1999 ....................................................

PAGE

9

4-2 Annual occupational environmental radionuclide inhalation, Area IV...................................... 11

4-3 Annual occupational environmental radionuclide inhalation at De Soto Avenue, 

Vanowen Building, and Downey facilities. .............................................................................. 12

4-4 External occupational environmental dose ............................................................................. 15

4-5 Maximum site-wide median intakes via inadvertent soil ingestion .......................................... 17
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 4 of 21

ACRONYMS AND ABBREVIATIONS

AERD Atomic Energy Research Department

AI Atomics International

Bq becquerel

Ci curie

cm centimeter

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ETEC Energy Technology Engineering Center

GIF Gamma Irradiation Facility

hr hour

m meter

mg milligram

mrem millirem

NAA North American Aviation

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

RIHL Rockwell International Hot Laboratory (later simply Hot Laboratory)

RMDF Radioactive Material Disposal Facility

RMHF Radiation Materials Handling Facility

s second

SNAP Systems for Nuclear Auxiliary Power

SRE Sodium Reactor Experiment

SSFL Santa Susana Field Laboratory

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

W watt

WBNS Water Boiler Neutron Source

yr year

§ section
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4.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health (NIOSH) in the completion of the individual work required for each dose 

reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the [probability of causation] guidelines established under subsection 

(c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation guidelines (nor the 

dose reconstruction regulation) define “performance of duty” for DOE employees with a covered 

cancer or restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

This TBD is part of the overall ETEC Site Profile, which describes plant facilities and processes, 

historical information, and environmental data in relation to dose reconstruction for ETEC workers. It 

contains supporting documentation to assist in the evaluation of occupational environmental doses 

from these processes. 
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 6 of 21

4.2 OPERATIONS OVERVIEW 

Occupational environmental dose refers to exposures received by workers while on the site but 

outside the facilities at the Energy Technology Engineering Center (ETEC) from facility discharges to 

the atmosphere, from ambient external radiation originating in the facilities, and from inadvertent 

ingestion of site-generated radionuclides. This TBD describes the estimated annual intakes for 

inhalation exposure, estimated radiation dose from ambient external exposures, and estimated 

inadvertent ingestions at ETEC. Environmental measurements do not distinguish the source of 

emissions and therefore reflect air concentrations from nearby as well as distant sources. The 

estimates of emissions were useful in filling in some gaps in measurement data and were critical to 

estimating exposures before the start of comprehensive and routine measurement data reports.

The receptors of concern are ETEC employees who did not wear external dosimetry or who were not 

monitored for internal exposures. To provide the basis for estimating the environmental dose for 

years when monitoring did not occur or was not sufficient to apply to coworkers, this TBD provides 

annual intakes and ambient external dose from 1952 to 1999.

ETEC, operated by The Boeing Company and its processors, played an important role in the 

development of the U.S. nuclear program. ETEC operations involved research, development, and 

testing of nuclear reactors, nuclear support operations, and non-nuclear energy research and 

development (Sapere and Boeing 2005, p. 2-2). 

From 1948 to 1955, the Atomic Energy Research Department (AERD) of North American Aviation, 

Inc. (NAA) occupied a portion of Building 001 at the Downey Plant on Lakewood Boulevard in 

Downey, California. On April 21, 1952, a small aqueous homogeneous reactor was constructed in an 

area that is now a loading dock; an environmental monitoring program was apparently initiated during 

1952 as well. The reactor and its associated exponential pile were apparently used as a source of 

neutrons for reactor physics experiments. The Water Boiler Neutron Source (WBNS) used a 93% 

enriched uranyl sulfate solution and operated at power levels up to 4 W at Downey until December 

1955. In 1956 it was dismantled and moved to a facility at the Santa Susana Field Laboratory (SSFL). 

Little information is available about the specific uses of the area after the relocation. All available 

records indicate that the plant was not left in a contaminated condition, and this was confirmed by a 

survey in 2000 (Liddy and Rutherford 2001). Other radioisotopes used at the Downey facility were 

apparently for industrial radiography and were not likely a source of internal exposure.

In 1956, NAA formed Atomics International (AI) as one of its divisions, replacing AERD. Between 

1956 and 1960, AI performed nuclear research and development at a facility known as the Vanowen 

Building (Building 038) on Canoga Avenue in Canoga Park, California. AI designed, developed, and 

operated two small aqueous 93% enriched uranyl sulfate research reactors designated L-47 and L-77 

at the Vanowen facility. The maximum power ratings were 5 and 10 W, respectively. Other 

operations included reactor design, fuel development, and radiochemistry. L-47 operated between 

August 1957 and June 1958, and L-77 operated between May 1958 and February 1960. The reactors 

were in the southeast corner of the building in what is now a loading dock. Historical information 

indicates that the areas that supported nuclear operations were routinely surveyed up to and following 

the removal of all radioactive material. No contamination above the current limits for unrestricted use 

was found. This was confirmed by a 1995 survey of the facility by the U.S. Nuclear Regulatory 

Commission (NRC) (Rutherford 2002). Subsequent occupants of the building would be unlikely to be 

exposed to contamination levels of importance for internal dose reconstruction.

In 1959, AI moved to a new facility on De Soto Avenue in Canoga Park. All radiological activities from 

the Vanowen Building, including the L-77 reactor, were transferred to the new headquarters facility. 
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 7 of 21

Until 1983, nuclear operations were conducted at the De Soto facility in Buildings 101 and 104 (until 

1984 these buildings were known as 001 and 004, respectively). Work continued at a much-reduced 

level in Building 104 until the mid-1990s. The L-77 reactor in Room 416-61 of Building 104 was 

decommissioned and decontaminated in the late 1970s. The facility was released for unrestricted use 

in February 1982. Many nuclear fuel manufacturing operations were conducted in the 1960s and 

1970s in both Building 101 and Building 104. These operations used 2% to 93% enriched uranium 

metal and composites. 

Two other operations in Building 104 at the De Soto facility used radioactive material. A Gamma 

Irradiation Facility (GIF) used sealed 137Cs and 60Co sources for hardness testing and food irradiation 

research. Biannual leak tests of these sources confirmed that internal exposures from these devices 

were unlikely. Operations ceased in the late 1980s; and, following removal of all equipment, the State 

of California released the GIF for unrestricted use in July 1999. The other Building 104 operation, the 

Mass Spectroscopy Laboratory or Helium Lab, analyzed radioactive samples for helium content. 

Operations ceased in 1995 and the mass spectroscopy equipment was sent to Richland, Washington. 

In 1998, all remaining support equipment was removed and the facility was decontaminated. In 

October 1998, the State released the Helium Lab for unrestricted use. In all, 13 separate radiation 

surveys were conducted in Buildings 101 and 104 to demonstrate that no residual contamination 

existed that would be of interest for dose reconstruction (Boeing 2003).

SSFL consisted of four administrative areas as well as a buffer zone. Area IV was established by 

NAA at SSFL in 1953 as a nuclear research and development facility1

For simplicity, the remainder of this document uses ETEC to refer to all of the historical organizations 

that operated at the De Soto Avenue and Vanowen Building facilities in Canoga Park, Downey, and 

Area IV of the SSFL. Activities conducted at all of these ETEC sites and facilities are evaluated in this 

report. 

. Starting in 1954 several 

nuclear reactors and critical assemblies were built and operated in Area IV. The Agency for Toxic 

Substances and Disease Registry (ATSDR) investigated radiological incidents occurring at SSFL

(ATSDR 1999) and stated: “Of all these incidents, only the Sodium Reactor Experiment (SRE) Fuel 

Damage incident, commonly known as "The Meltdown," resulted in a measurable release of 

radioactive material into the environment.” The SRE was a sodium-cooled graphite-moderated 

reactor in Building 143, which in 1959 had a loss of coolant that resulted in damage to 13 fuel 

assemblies (Sapere and Boeing 2005, pp. 2-1, 2-3, 2-5). Radioactive noble gases reached the outer 

containment and 28 Ci of radioactive noble gases 133Xe and 85Kr were released in a controlled manner 

and low concentrations to limit the potential dose (Rutherford 2005). A government-owned contractoroperated organization was formed to conduct non-nuclear research at the site. It was renamed ETEC 

in 1978. Most nuclear research programs and operations ceased in 1988. All research ended in 

1998. ETEC was given the job of decontaminating and decommissioning the former nuclear facilities. 

Many of the facilities used in Area IV have been decontaminated and decommissioned, but some of 

the work is ongoing.

4.3 INHALATION OF ONSITE AIRBORNE RADIONUCLIDES 

The environmental monitoring program at Area IV was established in May 1954 before construction of 

the first radiological facility (Sapere and Boeing 2005, p. 3-13), with emphasis on soil, vegetation, and 

water sampling in the environment around Area IV. The program had been initiated in 1952 at the 

Downey Facility but was terminated in the move to Canoga Park (Moore, Fisher, and Rowe 1962). 

 

1 More recently, buildings in Area IV have had the number 4 added to the beginning of the building designation. 

For example, the original Building 055 would now be referred to as Building 4055.
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From 1959 to present, ambient gross beta activity in air has been measured continuously in five 

locations. From 1963 on, gross alpha activity was measured. Ambient air samples are currently 

analyzed for isotopic content (Sapere and Boeing 2005, p. 3-15). No information is available for the 

Downey or Canoga Park facilities other than the De Soto facilities; information for the De Soto 

facilities is not available until ambient data are reported together with Area IV data in environmental 

monitoring reports. The ETEC ambient air monitoring data through 1999 are reported as gross alpha 

and gross beta without identification of specific radionuclides. In all cases the gross alpha and gross 

beta concentrations are a small fraction of the applicable limits and are indistinguishable from 

background. Because of this, stack emission data were used as the primary basis for inhalation 

intake estimates from onsite atmospheric radionuclide concentrations. 

Average annual gross alpha/gross beta concentrations in facility stack emissions were the basis for 

estimating potential worker environmental inhalation intakes. Most of the available ETEC stack 

emission data include annual average gross alpha and gross beta concentrations at the stack point of 

release. Years with data vary by facility, but gross alpha/gross beta concentration information is 

available for most years between 1971 and 1999.. For years where there were no emission data, the 

facility-specific average for available years was used. The average of yearly data was used rather 

than the median because the average concentration was higher. In many cases the stack emission

concentrations of gross alpha and gross beta are indistinguishable from the ambient background 

concentrations, indicating manmade radionuclide releases were very low. 

Identification of specific radionuclides released from various facilities in stack emissions are available 

in data from 1988 to 1999 and were used to characterize radionuclide emissions for all years. There 

are no radionuclide-specific data from earlier years. The radionuclide-specific data showed that 

naturally occurring radionuclides were major constituents of radionuclide emissions, with 210Po a main

constituent of the gross alpha concentration and 40K and 7

Be major constituents of the gross beta 

detected. Any uranium and thorium detected is assumed to originate from facility activities, even 

though it could be naturally occurring.

4.3.1

Radionuclide-specific, gross alpha and gross beta stack emission data are available for this period for 

some facilities. In Area IV these facilities include mainly Building 020

Stack Effluent Concentrations, 1988 to 1999

2

For De Soto the only facility for which there is emission data is Building 004/104 (Hot Analytical 

Chemistry). The most complete information, and the majority of the radioactive releases, comes from 

Building 020 for Area IV. These effluent data show that the majority of the release during most years 

was from naturally occurring radionuclides: 210Po for alpha-emitters and 40K and 7

Be for beta-emitters. 

No data were available for 1993 and 1994. 

 [RIHL, Rockwell International 

Hot Laboratory, later simply Hot Laboratory) and Buildings 021-022 [Radioactive Materials Disposal

Facility (RMDF), Radioactive Materials Handling Facility (RMHF)], with limited information for T024 

[Systems for Nuclear Auxiliary Power (SNAP) Environmental Test Facility], and T059 (SNAP 8 

Development Reactor) (Moore 1989; Robinson 1998; Gaylord 1990; Rutherford 1999; Rockwell 1991, 

1992, 1993, Rockwell 1996; Boeing 2000; Tuttle 1997). 

Table 4-1 lists the radionuclides identified as being released from each facility during this period. 

Naturally occurring radionuclides are not included; isotopes of thorium and uranium could be naturally 

occurring but are not so considered in ETEC effluent reports. 

 

2

As noted in an earlier footnote, the original Building 020 may be referred to as Building 4020.
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Table 4-1. Identified radionuclides in facility air effluents, 1988 to 1999. 

Facility Alpha emitters Beta emitters

020, Hot Lab Pu-238, Pu-239, U-234, U-235, U-238, Th-228, Th-230, 

Th-232, Am-241

Co-60, Sr-90, Cs-137

021-022, RMDF/RMHF Pu-238, Pu-239, U-234, U-235, U-238, Th-228, Th-230, 

Th-232, Am -241

Co-60, Sr-90, Cs-137, 

H-3

T024, SNAP ETF Th-230 Cs-137

T059, SNAP8DR Pu-238, Pu-239, U-234, U-238, Am-241 Co-60, Cs-137

004/104, De Soto Pu-238, Pu-239, U-234, U-235, U-238, Th-228, Th-230, 

Th-232, Am -241

Co-60, Sr-90, Cs-137, 

Pu-241

4.3.2

Gross alpha and gross beta effluent data are available for most years during this period for Area IV 

facilities Buildings 020, 021-022, and Building 055 (Nuclear Materials Development Facility), and 

limited data is available for Building 003 (Engineering Test Building) (Remley 1979, Eggleston 1983, 

1984; Hart 1979, 1980a,b,c,d; Moore c. 1974; Heine 1973, Oldenkamp and Mills 1991; Tuttle 1985, 

1986a,b,c, 1988a,b]. 

Stack Effluent Concentrations, 1971 to 1987

Data are available for De Soto Building 004/104 and 001/101 (Fuel Fabrication). Because of lack of 

other information, Building 001/101 was assumed to have the same radionuclide composition in 

effluents as Building 004/104. This is a conservative assumption because, as a fuel fabrication 

facility, no mixed fission products would be expected in the gross beta emissions. In years where 

data were not available, stack concentrations were assumed to be the average yearly gross alpha and 

gross beta concentrations in stack effluents from years 1971 to 1999 for which data were available. 

4.3.3

No data were available for this period, but because operations and activities during this period took 

place in many of the same facilities addressed in later years, the average yearly gross alpha and 

gross beta concentrations in stack effluents from 1971 to 1999 were assumed for each of these years. 

Furthermore, the average percentage that each identified radionuclide contributed to the gross alpha 

or gross beta concentration determined from 1988 to 1999 data was applied to each of these years to 

make radionuclide-specific stack concentration estimates. Because no radionuclide-specific data 

were available for Building 055, the characteristic effluent radionuclide percentage from Buildings 024 

and 059 were applied. Radionuclide-specific data were also unavailable for De Soto Building 

001/101, so the characteristic effluent radionuclides percentages from Building 004/104 were applied. 

Stack Effluent Concentrations, 1959 to 1970

4.3.4

No data are available for the period before 1959; the facilities operating during this period were mainly 

small reactor facilities. The SRE, the largest reactor at ETEC, operated from 1954 to 1964 in Area IV. 

It is possible that reactor operations before 1959 could have released fission product activities similar 

to those after 1959. Therefore, fission product releases from Area IV from 1954 to 1958 are assumed 

to be the same as 1959 to 1964. There would have been few activities or operations with potential for 

release of alpha-emitting radionuclides; therefore, alpha-emitting radionuclides from 1954 to 1958 are 

assumed to be 10 percent of alpha-emitting radionuclide releases from 1959 to 1964. Operations at 

the De Soto Avenue facilities did not begin until 1959. Activities at the Downey site from 1952 to 1955

and Canoga Park Vanowen Building from 1954 to 1960 were limited and releases are conservatively 

assumed to be the same as the 1954 Area IV releases. 

Stack Effluent Concentrations, Before 1959
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 10 of 21

4.3.5

The average annual stack effluent concentrations described in Section 4.3.3 were used as the basis 

for estimating occupational environmental inhalation intakes from facility effluent releases. The facility 

effluent concentrations for each year were summed. Workers were assumed to be exposed for 

2,000 hr/yr with a breathing rate of 3.3 × 10-4 m3

/s (ICRP 1975). Because the stack effluent 

concentrations were at the point of release, a further reduction factor of 0.01 was taken to account for 

the lessened overall intake due to contribution from multiple, widely spaced facilities; atmospheric 

dispersion of stack effluent over the course of a year’s exposure, and building wake effects.

Estimates of Potential Inhalation Intake, 1952 to 1999

Noble gases are not included in the estimates of intake because they are not included in emission 

monitoring data and because the potential dose from radioactive noble gases is very small. In 

addition and with one exception, acute release events that contribute negligibly to the annual average 

radionuclide concentration are not included. The exception is the inclusion of mixed fission product 

releases from a burned fuel slug event occurring in Building 020 in July 1975, which had a noticeable 

effect on the average annual gross beta concentrations (Hart 1979). 

Table 4-2 lists the estimated environmental inhalation intakes in Area IV. Table 4-3 lists the estimated 

environmental inhalation intakes at the Canoga Park De Soto Avenue and Vanowen Building facilities

and the Downey site. In the absence of measurements or studies, NIOSH guidance requires that 

default solubility classes and particle size values from the International Commission on Radiological 

Protection be used (NIOSH 2002, pp. 15–16). Facility-specific solubility and particle size data are not 

available. The most claimant favorable material form and solubility should be selected for each 

radionuclide at all sites and facilities. 
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 11 of 21

Table 4-2. Annual occupational environmental radionuclide inhalation, Area IV (Bq/yr). 

Year

Radionuclide 

Pu239Pu238Am241 U-234 U-235 U-238Th228Th232Th230 Co-60 Sr-90 Cs-137 H-3

1999 1.1E-04 2.4E-06 6.4E-06 2.0E-04 4.1E-06 4.7E-04 8.0E-06 9.1E-06 1.8E-03 1.1E-02 3.0E-02 9.0E-02 2.1E-02

1998 5.3E-04 4.2E-05 1.4E-05 5.7E-04 3.0E-05 7.7E-04 5.8E-05 1.2E-04 4.3E-03 3.5E-02 3.1E-02 1.4E-01 2.3E-01

1997 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 4.9E-03 1.1E-03 8.1E-03 2.9E-01

1996 6.5E-05 2.0E-06 5.7E-06 1.3E-04 4.9E-06 4.9E-04 9.6E-06 9.2E-06 6.1E-04 6.3E-03 1.3E-03 1.4E-02 1.9E-01

1995 1.9E-04 1.0E-05 1.6E-05 3.3E-04 1.8E-05 1.5E-03 3.5E-05 4.0E-05 6.3E-04 8.3E-03 2.7E-03 2.0E-02 2.9E-01

1994 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1993 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1992 9.8E-05 7.9E-06 6.0E-06 1.3E-04 9.2E-06 5.2E-04 1.8E-05 2.6E-05 1.6E-04 1.9E-03 7.0E-04 4.1E-03 9.3E-02

1991 2.1E-04 1.7E-05 1.0E-05 2.6E-04 1.7E-05 8.0E-04 3.3E-05 5.2E-05 8.9E-04 5.4E-03 1.5E-03 1.3E-02 1.7E-01

1990 1.4E-04 7.6E-06 1.1E-05 2.3E-04 1.2E-05 9.7E-04 2.5E-05 2.9E-05 6.3E-04 4.4E-03 8.5E-04 1.1E-02 7.4E-02

1989 3.3E-05 1.1E-06 1.1E-06 5.2E-05 6.6E-07 5.5E-05 1.3E-06 2.8E-06 6.1E-04 5.6E-03 5.1E-04 1.1E-02 1.5E-01

1988 1.2E-04 1.0E-05 3.6E-06 1.2E-04 8.1E-06 2.5E-04 1.6E-05 3.0E-05 6.5E-04 7.2E-03 1.2E-03 1.5E-02 2.4E-01

1987 6.5E-05 3.9E-06 2.9E-06 9.1E-05 3.8E-06 2.1E-04 7.4E-06 1.2E-05 6.2E-04 1.5E-02 2.5E-03 2.7E-02 7.2E-01

1986 5.6E-05 5.2E-06 1.9E-06 5.6E-05 4.3E-06 1.4E-04 8.4E-06 1.5E-05 1.8E-04 1.5E-02 4.2E-03 3.4E-02 5.6E-01

1985 1.2E-04 8.2E-06 9.2E-06 1.8E-04 1.2E-05 8.2E-04 2.3E-05 2.9E-05 1.3E-04 1.5E-02 2.1E-02 7.0E-02 5.5E-01

1984 7.7E-05 8.0E-06 2.3E-06 6.7E-05 6.3E-06 1.7E-04 1.2E-05 2.3E-05 1.0E-04 1.0E-03 8.2E-04 3.9E-03 1.6E-02

1983 4.9E-05 4.9E-06 1.0E-06 4.2E-05 3.4E-06 6.1E-05 6.7E-06 1.4E-05 1.9E-04 1.5E-03 3.7E-04 3.5E-03 4.8E-02

1982 2.0E-04 2.3E-05 2.4E-06 1.2E-04 1.5E-05 1.0E-04 3.0E-05 6.4E-05 1.4E-04 1.8E-03 2.7E-03 9.6E-03 2.5E-02

1981 8.4E-05 8.9E-06 2.0E-06 6.9E-05 6.6E-06 1.3E-04 1.3E-05 2.5E-05 1.7E-04 4.9E-03 3.0E-03 1.5E-02 1.7E-01

1980 2.1E-04 1.4E-05 6.9E-06 2.7E-04 1.0E-05 4.3E-04 2.0E-05 3.8E-05 2.3E-03 9.6E-03 3.2E-03 2.5E-02 2.3E-01

1979 1.3E-04 1.3E-05 3.3E-06 1.0E-04 1.0E-05 2.3E-04 2.0E-05 3.8E-05 1.9E-04 5.3E-03 6.2E-03 2.1E-02 2.3E-01

1978 1.1E-04 1.1E-05 3.0E-06 9.3E-05 8.4E-06 2.2E-04 1.7E-05 3.1E-05 2.0E-04 1.3E-02 9.2E-03 4.1E-02 4.9E-01

1977 1.3E-04 1.4E-05 2.2E-06 8.9E-05 9.6E-06 1.2E-04 1.9E-05 3.8E-05 2.1E-04 5.9E-03 1.7E-03 1.5E-02 1.6E-01

1976 2.4E-04 2.8E-05 3.7E-06 1.6E-04 1.9E-05 1.9E-04 3.7E-05 7.7E-05 2.7E-04 4.0E-03 6.9E-04 1.0E-02 5.1E-02

1975 4.7E-04 5.5E-05 5.9E-06 2.9E-04 3.6E-05 2.6E-04 7.1E-05 1.5E-04 4.1E-04 1.6E-01 6.1E-01 1.7E+00 5.1E-01

1974 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1973 1.6E-04 1.7E-05 3.8E-06 1.3E-04 1.3E-05 2.6E-04 2.5E-05 4.9E-05 2.1E-04 3.9E-03 1.3E-03 7.5E-03 2.3E-01

1972 1.6E-04 1.6E-05 4.8E-06 1.4E-04 1.3E-05 3.6E-04 2.6E-05 4.7E-05 2.2E-04 1.0E-02 3.0E-02 8.7E-02 6.2E-02

1971 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1970 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1969 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1968 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1967 1.6E-04 1.4E-05 5.4E-06 1.6E-04 1.1E-05 3.9E-04 2.2E-05 4.0E-05 6.6E-04 1.4E-02 3.1E-02 9.5E-02 2.3E-01

1966 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1965 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1964 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1963 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1962 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1961 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1960 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1959 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1958 1.3E-05 1.4E-06 4.4E-07 1.2E-05 1.1E-06 3.4E-05 2.2E-06 4.0E-06 5.1E-06 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1957 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1956 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1955 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1954 1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 12 of 21

Table 4-3. Annual occupational environmental radionuclide inhalation at De Soto Avenue, Vanowen 

Building, and Downey facilities. (Bq/yr). 

Year

Radionuclide 

Pu239Pu238Am241 U-234 U-235 U-238Th228Th232Th230 Co-60 Sr-90Cs137 Pu-241

1954-1999, De Soto Avenue

1999 0 0 0 0 0 0 0 0 0 0 0 0 0

1998 9.1E-07 1.1E-05 2.6E-05 2.1E-04 1.4E-05 2.6E-05 1.1E-04 4.0E-05 4.5E-05 3.9E-05 3.6E-04 1.8E-04 0

1997 3.7E-06 4.3E-05 1.1E-04 8.6E-04 5.6E-05 1.1E-04 4.4E-04 1.6E-04 1.8E-04 3.4E-05 3.1E-04 1.5E-04 0

1996 1.8E-07 2.1E-06 5.0E-06 4.1E-05 2.7E-06 5.0E-06 2.1E-05 7.8E-06 8.7E-06 1.1E-05 1.0E-04 4.9E-05 0 

1995 3.6E-07 4.2E-06 1.0E-05 8.4E-05 5.5E-06 1.0E-05 4.3E-05 1.6E-05 1.8E-05 1.9E-05 1.7E-04 8.4E-05 6.5E-03

1994 7.6E-06 8.9E-05 2.2E-04 1.8E-03 1.2E-04 2.2E-04 9.1E-04 3.3E-04 3.7E-04 5.1E-05 4.8E-04 2.3E-04 0

1993 7.6E-06 8.9E-05 2.2E-04 1.8E-03 1.2E-04 2.2E-04 9.1E-04 3.3E-04 3.7E-04 5.1E-05 4.8E-04 2.3E-04 0 

1992 1.4E-06 1.6E-05 4.0E-05 3.2E-04 2.1E-05 3.9E-05 1.7E-04 6.1E-05 6.8E-05 2.6E-05 2.4E-04 1.2E-04 0

1991 1.2E-06 1.4E-05 3.4E-05 2.8E-04 1.8E-05 3.4E-05 1.4E-04 5.3E-05 5.9E-05 4.9E-05 4.6E-04 2.2E-04 0

1990 1.8E-06 2.1E-05 5.1E-05 4.2E-04 2.7E-05 5.1E-05 2.1E-04 7.9E-05 8.8E-05 9.5E-05 8.9E-04 4.3E-04 0

1989 7.6E-06 8.9E-05 2.2E-04 1.8E-03 1.2E-04 2.2E-04 9.1E-04 3.3E-04 3.7E-04 5.1E-05 4.8E-04 2.3E-04 0

1988 4.2E-06 4.9E-05 1.2E-04 9.8E-04 6.4E-05 1.2E-04 5.0E-04 1.9E-04 2.1E-04 6.8E-05 6.4E-04 3.1E-04 0

1987 5.2E-06 6.1E-05 1.5E-04 1.2E-03 7.9E-05 1.5E-04 6.2E-04 2.3E-04 2.6E-04 4.9E-05 4.6E-04 2.2E-04 0

1986 1.4E-06 1.6E-05 4.0E-05 3.2E-04 2.1E-05 4.0E-05 1.7E-04 6.2E-05 6.9E-05 5.8E-05 5.4E-04 2.6E-04 0 

1985 3.3E-06 3.9E-05 9.5E-05 7.7E-04 5.1E-05 9.5E-05 4.0E-04 1.5E-04 1.6E-04 4.2E-05 3.9E-04 1.9E-04 0

1984 1.0E-05 1.2E-04 3.0E-04 2.4E-03 1.6E-04 2.9E-04 1.2E-03 4.5E-04 5.1E-04 5.5E-05 5.1E-04 2.5E-04 0

1983 4.5E-04 5.3E-03 1.3E-02 1.0E-01 6.8E-03 1.3E-02 5.4E-02 2.0E-02 2.2E-02 6.6E-04 6.2E-03 3.0E-03 0

1982 1.0E-05 1.2E-04 2.9E-04 2.3E-03 1.5E-04 2.9E-04 1.2E-03 4.5E-04 5.0E-04 4.7E-05 4.4E-04 2.2E-04 0

1981 2.2E-05 2.6E-04 6.3E-04 5.1E-03 3.3E-04 6.3E-04 2.6E-03 9.7E-04 1.1E-03 1.4E-04 1.3E-03 6.3E-04 0

1980 3.9E-06 4.6E-05 1.1E-04 9.1E-04 6.0E-05 1.1E-04 4.7E-04 1.7E-04 1.9E-04 1.9E-04 1.8E-03 8.6E-04 0

1979 1.8E-04 2.1E-03 5.0E-03 4.1E-02 2.7E-03 5.0E-03 2.1E-02 7.8E-03 8.7E-03 2.0E-04 1.8E-03 8.9E-04 0

1978 2.6E-06 3.0E-05 7.5E-05 6.0E-04 4.0E-05 7.4E-05 3.1E-04 1.1E-04 1.3E-04 1.3E-04 1.2E-03 5.7E-04 0

1977 2.2E-06 2.6E-05 6.2E-05 5.1E-04 3.3E-05 6.2E-05 2.6E-04 9.6E-05 1.1E-04 1.2E-04 1.1E-03 5.3E-04 0

1976 3.0E-04 3.5E-03 8.5E-03 6.9E-02 4.5E-03 8.5E-03 3.6E-02 1.3E-02 1.5E-02 3.6E-04 3.4E-03 1.6E-03 0 

1975 2.4E-05 2.8E-04 6.9E-04 5.6E-03 3.7E-04 6.9E-04 2.9E-03 1.1E-03 1.2E-03 1.2E-04 1.1E-03 5.4E-04 0

1974 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1973 9.7E-06 1.1E-04 2.8E-04 2.2E-03 1.5E-04 2.8E-04 1.2E-03 4.3E-04 4.8E-04 7.7E-05 7.2E-04 3.5E-04 0

1972 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1971 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1970 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1969 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1968 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1967 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1966 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0 

1965 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1964 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1963 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0 

1962 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1961 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1960 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1959 9.4E-05 1.1E-03 2.7E-03 2.2E-02 1.4E-03 2.7E-03 1.1E-02 4.1E-03 4.6E-03 1.9E-04 1.8E-03 8.9E-04 0

1954-1960, Vanowen Building (all years are the same)

1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

1952-1955, Downey Site (all years are the same)

1.3E-04 1.4E-05 4.4E-06 1.2E-04 1.1E-05 3.4E-04 2.2E-05 4.0E-05 5.1E-05 1.1E-02 3.1E-02 8.8E-02 2.3E-01

4.4 AMBIENT EXTERNAL RADIATION

Ambient external radiation dose information for ETEC is available for 1975 to 1999. During the later 

years (1974 to 1989) of the active nuclear programs at ETEC, the ambient radiation monitoring 

program used complex bulb dosimeters (CaF2:Mn). This was justified by the amount of nuclear 

materials handled in the operations in Area IV and De Soto and by the low levels of radiation in the 

environment. At the termination of all nuclear work in 1989, that level of monitoring was no longer 

needed and efforts were directed toward simplifying the program. The initial step included using the 

well-established LiF thermoluminescent dosimeters (TLDs) for personnel monitoring in radiation work. 
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 13 of 21

While these dosimeters were well suited to measuring exposures in the range of interest for 

compliance with occupational radiation regulations (doses above background), they were somewhat 

insensitive for environmental measurements because the resolution in terms of dose uses increments 

of 10 mrem/quarter. Using these dosimeters demonstrated that environmental exposures did not 

reach regulatory limits, but it provided limited information on the actual exposure rates around the 

facilities and in the neighboring environment (Rutherford 1999).

In addition to the LiF TLDs, Rocketdyne began deploying environmental TLDs in the last quarter of

1995 that used an aluminum oxide (called sapphire) chip. These TLDs were capable of determining 

doses in increments of 0.1 mrem (in comparison to 10 mrem for the LiF badges). Proper use of the 

control badges supplied with these dosimeters allowed elimination of the natural and transportation 

exposure that occurred before, during, and after the deployment of the environmental dosimeters to 

measure the ambient radiation. This permitted accurate determination of the net exposure received 

while the environmental TLDs were exposed in the field to the ambient radiation (Rutherford 1999).

The following sections describe estimates of environmental external dose during five periods from 

1954 to 1999; estimates for possible earlier exposures in Downey and Canoga Park are also 

provided. Workers were assumed to be exposed for 2,000 hr/yr. A baseline estimate of the 

environmental external dose was developed for each period, and additional dose for specific facilities 

during specific years was developed based on aerial survey information and specific facility fence line 

exposure rate information. In all cases, the minimum dose above background is 10 mrem/yr.

4.4.1

The sensitivity of the aluminum oxide chip in environmental TLDs is accounted for during this period. 

Average annual dose from offsite background dosimeters is subtracted from the average annual dose 

of environmental TLDs in Area IV and De Soto to provide an estimate of the baseline occupational 

environmental dose at the respective sites. The baseline occupational environmental dose for both 

De Soto and Area IV is 10 mrem/yr. In addition, environmental dosimeters near Buildings 021-022 

(RMDF/RMHF) are consistently higher than the average. This is stated to be reflective of normal 

operations at the RMHF (Rutherford 1999). Occupational environmental doses near this facility are

increased by 10 mrem/yr during this period to 20 mrem/yr. 

External Dose, 1996 to 1999

4.4.2

Relatively insensitive LiF chips were used in dosimeters during this period, so the baseline dose 

during this period was based on the minimum dosimeter sensitivity of 10 mrem/quarter or 40 mrem/yr. 

This minimum sensitivity baseline results in a higher annual dose than subtraction of offsite 

background dose from onsite environmental dosimeter results. The occupational environmental 

external dose near Buildings 021-022 is increased by 10 mrem/yr because of elevated dosimeter 

readings.

External Dose, 1990 to 1995

4.4.3

External doses during this period of active nuclear operations at ETEC were measured using the 

complicated CaF2:Mn bulb dosimeters. Baseline external occupational environmental doses during 

this period were estimated by subtracting the average annual offsite background dose from the 

maximum annual onsite dose at De Soto and Area IV. Baseline doses were at the minimum of 10 

mrem/yr except for 1978, when radiation from the Building 621 area, described below, was high 

enough that the baseline dose for the site was doubled to 20 mrem/yr. 

External Dose, 1975 to 1989
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Three documents provided quarterly facility fence line exposure rates for Building 20 (RIHL) and 

Buildings 021-022 (RMDF/RMHF) during 1986, 1987, and 1988 (Tuttle 1988a,b, 1989). These 

measurements showed the exposure rates around Building 020 to be background, but the exposure 

rates around the RMDF to be elevated above the baseline. These measurements were used as the 

basis for elevating external occupational environmental doses around the RMDF along with results of 

an aerial radiological survey in 1978 (Jobst 1979). The aerial survey represents a snapshot in time, 

but shows elevated exposure rates around the RMDF, around Building 143 the Sodium Reactor 

Experiment (SRE) Facility from storage of radioactive material, and around the Building 621 

Radioactive Accountable Waste Storage Building from storage of radioactive liquid in a nearby tank 

vault. Because of the nature of the survey, these elevated exposure rates were applied only to the 

year 1978 for the SRE facility. However, because elevated exposure around the RMDF had occurred 

in other years, the aerial survey results were included in determining an average elevated dose rate 

for the RMDF. Because Building 621 was part of the RMDF complex, the RMDF elevated results 

were applied throughout this period to Building 621 as well.

4.4.4

No data are available for the period from 1959 to 1974, but many of the same nuclear operations 

continued during this period. The facilities not considered are mainly small reactor facilities, which are 

typically well shielded. The SRE, the largest reactor at ETEC, operated from 1954 to 1964. 

Building 621 was constructed in the mid-1960s (Sapere and Boeing 2005, Volume 2, Part 5). Both 

Building 020, the Hot Lab, and Buildings 021-022, the RMDF, began operations in 1959. Based on 

operating information from 1975 to 1989, estimates were made of the occupational environmental 

external dose for 1959 to 1974. The baseline dose above background in Area IV was assumed to be 

doubled to 20 mrem/yr to account for reactor operations during this period. The baseline dose in 

De Soto remains at 10 mrem/yr. Facility-specific dose during these years was assumed to be 

elevated to 20 mrem/yr for De Soto Building 001 and elevated to 80 mrem/yr for the RMDF in Area IV. 

During the years of SRE operations, a facility-specific dose of 40 mrem/yr was assumed for Building 

143. 

External Dose, 1959 to 1974

4.4.5

No data are available for the period before 1959; as noted above the facilities were mainly small 

reactor facilities, which are typically well shielded. The SRE, the largest reactor at ETEC, operated

from 1954 to 1964 in Area IV. The baseline dose above background in Area IV was assumed to be 

20 mrem/yr to account for reactor operations during this period. The facility-specific dose for Building 

143 was 40 mrem/yr for 1954 to 1959. Environmental external doses from 1952 to 1955 for Downey 

and from 1954 to 1960 for the Canoga Park Vanowen Building are estimated to be 10 mrem/yr based 

on the operation of small test reactors. 

External Dose Before 1959

4.4.6

Table 4-4 lists the occupational environmental external dose estimates for Area IV, De Soto Avenue, 

Vanowen, and Downey sites for applicable periods from 1952 to 1999. Some facility-specific doses 

are provided for De Soto and in Area IV.

Estimates of External Dose, 1952 to 1999
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 15 of 21

Table 4-4. Annual external occupational environmental dose (mrem/yr). 

Year

De Soto Area IV

Baselinea 

Facilityspecific dose

Baselinea 

Facility-specific dose

Fuel Fab 

Bldg 001

RMDF 

Bldg 021-022

SRE 

Bldg 143

RAWSB 

Bldg 621

1999 -- -- 10 20

1998 10 10 20

1997 10 10 20

1996 10 10 20

1995 40 40 50

1994 40 40 50

1993 40 40 50

1992 40 40 50

1991 40 40 50

1990 40 40 50

1989 10 10 20 20

1988 10 10 40 40

1987 10 10 60 60

1986 10 10 160 160

1985 10 10 70 70

1984 10 10 70 70

1983 10 20 10 70 70

1982 10 20 10 70 70

1981 10 20 10 70 70

1980 10 20 10 70 70

1979 10 20 10 70 70

1978 10 20 20 40 90 180

1977 10 20 10 70 70

1976 10 20 10 70 70

1975 10 20 10 70 70

1974 10 20 20 80 80

1973 10 20 20 80 80

1972 10 20 20 80 80

1971 10 20 20 80 80

1970 10 20 20 80 80

1969 10 20 20 80 80

1968 10 20 20 80 80

1967 10 20 20 80 80

1966 10 20 20 80 80

1965 10 20 20 80

1964 10 20 20 80 40

1963 10 20 20 80 40

1962 10 20 20 80 40

1961 10 20 20 80 40

1960 10 20 20 80 40

1959 10 20 20 80 40

1958 -- 20 40

1957 -- 20 40

1956 -- 20 40

1955 -- 20 40

1954 -- 20 40

Vanowen

1954-1960 10

Downey

1952-1955 10

a. Minimum above-background baseline external dose of 10 mrem/yr.
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 16 of 21

4.5 INGESTION

Two sources of occupationally derived radioactive material for ingestion are possible:

• Potable water

• Removable surface contamination from floors, objects, or soils via the hand to the mouth

Potable water is not a source of occupational radioactive material at ETEC because the ETEC 

facilities used either bottled water from an offsite vendor (Moore, Fisher, and Rowe 1962) or the city 

water supply. SSFL continues to use bottled water. 

Ingestion of removable surface contamination is negligible because the intake is proportional to the 

amount of dirt consumed. In NRC NUREG/CR-5512, Kennedy et al. (1992, p. 6.14) assume the 

amount of soil ingested by an adult involved in building renovation work would be 10 mg/hr. More 

sedentary work would produce a smaller consumption, perhaps 50 mg/day. The other factor is the 

activity concentration in the ingested dirt. It is reasonable to expect that, outside of radiologically 

controlled areas, the removable surface contamination level or activity concentration in soil should on 

average approach background levels. The surface contamination level is unlikely to average an order 

of magnitude more than the background level for the entire workplace for the entire year.

Kennedy et al. (1992, Sections 3.2.5 and 6.3.2) define methods and assumptions for use in 

calculation of doses to individuals occupying a facility contaminated with residual radioactive material. 

An equation for calculation of ingestion dose is included. For purposes of EEIOCPA, dose is 

reconstructed by different algorithms than those used in this report, so only the ingestion intake is 

estimated here. 

The method equation can be summarized in a word description as:

Intake equals the exposure duration multiplied by the effective transfer rate of removable 

contamination from surface to hands then to mouth and multiplied by the average removable 

surface contamination level.

More specifically:

Ingestion t ××= ActGO

where

Ingestion = ingestion intake in Bq/yr

t = annual exposure duration (2,000 hr/yr)

GO = ingestion transfer rate of removable surface activity during building occupancy 

(1 × 10-4 m2/hr)

Act = removable surface contamination (activity) level (Bq/m2)

Therefore, full-time occupancy (2,000 hr/yr) in an area with a removable surface activity of a 

radionuclide at the level of 1 Bq/m2 would result in an intake of 0.2 Bq of that radionuclide per year. 

Inadequate data are available to estimate the median ingestion each year and at each facility, but

estimates or bounding assumptions are necessary for dose reconstruction. To conservatively bound 

ingestion from occupational environmental sources, a worker is assumed to be a full-time occupant 

(2,000 hr/yr) of an area contaminated with 239Pu and 90Sr at 100% of their current removable 

contamination limits. These NRC-defined contamination limits (AEC 1974) have been essentially 
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 17 of 21

unchanged for decades. The plutonium limit is 20 dpm/100 cm2 or 33 Bq/m2, the strontium limit is 

200 dpm/100 cm2 or 330 Bq/m2

. The resultant ingestion intake for all years and all ETEC facilities is 

bounded at a level of 6.7 Bq/yr of 239Pu and 67 Bq/yr of 90Sr, as shown in Table 4-5. The most 

claimant favorable material form and solubility should be selected for each radionuclide.

Table 4-5. Maximum site-wide median intakes (Bq/yr) via inadvertent soil ingestion. 

Site Years Radionuclidea

Pu-239 Sr-90

Area IV 1954 - 1999 6.7 67

De Soto 1959 - 1999 6.7 67

Vanowen 1954- 1960 - 67

Downey 1952 - 1956 - 67

a. Radionuclides assumed for consistency with the predominant ones used for internal dose calculation. Pu-239 is 

representative of alpha-emitting radionuclides, Sr-90 representative of beta/gamma-emitting radionuclides.

4.6 UNCERTAINTY

Uncertainty is lower in more recent years and higher in earlier years when less data are available. 

There are few available documents with estimates of uncertainty for ETEC environmental 

measurements. No information is available on uncertainty of stack effluent concentrations. Many 

results are reported as less-than values, which indicates concentrations below the minimum 

detectable concentration. Qualitative estimates of inhalation intakes are 50% at 1 standard deviation 

from 1988 to 1999, 100% from 1971 to 1987, and 200% from 1959 to 1970.

Some of the best estimates of uncertainty are associated with the later years for external dose. 

Qualitative estimates of occupational environmental external dose are 20% at 1 standard deviation 

from 1996 to 1999; 50% from 1990 to 1995; 100% from 1975 to 1989; and 200% from 1959 to 1974.

Qualitative estimate of ingestion intakes are 100% at 1 standard deviation.
Document No. ORAUT-TKBS-0038-4 Revision No. 00 Effective Date: 03/01/2006 Page 18 of 21

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GLOSSARY

alpha radiation 

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2. Alpha particles have low penetrating power and a short range (a few 

centimeters in air). Outside the body, the most energetic alpha particle generally fails to 

penetrate the dead layers of cells covering the skin or a sheet of paper. Alpha particles 

emitted by radionuclides inside the body are a more significant health risk. 

becquerel (Bq) 

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) Bq. 

beta radiation 

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

dosimetry 

Measurement and calculation of internal and external radiation doses.

exposure 

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

radioactivity 

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. 

rem 

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural. 

thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose.